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The following information was provided by the licensee via phone and email: On October 12, 2025, Columbia Generating Station (CGS) was performing a reactor building emergency cooling test. At 1428 (PDT), an air damper associated with the division 1 motor control center (MCC) room cooling failed to perform its intended function to close. This MCC supports the operation of one subsystem of the standby gas treatment system (SGT). At 1440, a second air damper, associated with the division 2 MCC room cooling, also failed. This MCC supports the operation of the other SGT subsystem. Field operators were dispatched to investigate the potential cause. At 1627, the first air damper that had failed was observed to have closed on its own without further operator action. Operators subsequently determined that failure of the air damper to close rendered the associated emergency room coolers inoperable. At 1705, (the division 2 MCC) was declared inoperable and technical specification action statement 3.8.7.A was entered. From 1440 to 1627, CGS was in a condition that required both SGT subsystems to be declared inoperable due to the loss of emergency room cooling to their associated MCCs. This condition constitutes a loss of safety function of SGT and secondary containment. It could have challenged the station's ability to control a radioactive release had one occurred during that time. This notification is being made pursuant to 10 CFR 50.72(b)(3)(v)(C) and (D) due to the loss of safety function of both trains of SGT and secondary containment for approximately 2 hours. The Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At the time of notification, the licensee had exited all technical specification action statements.

• * * RETRACTION ON 11/5/2025 AT 1925 EST FROM JERRY AINSWORTH TO ERNEST WEST * * *

On October 12, 2025, at 2152 PDT, Columbia Generating Station notified the NRC pursuant to 10 CFR 50. 72(b)(3)(v)(C) and (D) due to the loss of safety function of both trains of standby gas treatment (SGT) and secondary containment for approximately 2 hours under Event Notification 57982. The notification was made due to the failure of two dampers to close during testing of the Reactor Building Emergency Cooling System. The dampers that failed to close were cross-divisional and affected rooms with equipment necessary for both trains of the SGT system which supports the safety function of secondary containment. Following the event, engineering conducted an evaluation to determine if both air dampers remaining open during a loss of cooling accident would have put any safety-related equipment at risk due to potential changes in radiological and environmental conditions within the associated motor control center (MCC) rooms. The engineering evaluation determined temperatures and humidities would not have exceeded limits in the MCC rooms due to the dampers being left open. Additionally, any increase in radiation would be considered negligible and would not affect the ability of the equipment to perform their design functions. Therefore, the safety-related equipment within the MCC rooms would have been available and operable to perform their design function were an accident to occur. Consequently, the failure of two dampers to close during testing is not considered to be an event or condition that could have prevented fulfillment of the safety function of SGT and secondary containment and did not impact the ability to mitigate the consequences of an accident or control the release of radioactive material, therefore, event notification 57982 is retracted. The Resident Inspector has been notified. Notified R4DO (Vossmar)

The following information was provided by the licensee via email: During thermography of a reactor protection system (RPS) distribution panel, a circuit breaker (RPS-CB-7B) was inadvertently opened. This resulted in a partial loss of power to RPS Division B, which caused containment isolations to occur in multiple systems (Reactor Water Clean Up, Equipment Drains Radioactive, Floor Drains Radioactive, Reactor Recirculation, and Traversing lncore Probe). Specifically, RWCU-V-1, FDR-V-3, EDR-V-19, RRC-V-19, and TIP-V-15 all closed. All actuations occurred as designed upon the partial loss of RPS power. This event is being reported pursuant to 10 CFR 50.72(b)(3)(iv)(A) due to an unplanned valid actuation of a system pursuant to 10 CFR 50.72(b)(3)(iv)(B)(2). Additionally, this is being reported pursuant to 10 CFR 50.72 (b)(3)(xiii) for a major loss of emergency assessment capability due to the inability to assess primary containment identified and unidentified leakage rates. Emergency assessment capability was restored at 1008 PDT upon system restoration. The NRC resident was notified by the licensee.

• * * UPDATE FROM SIMEON MORALES TO DONALD NORWOOD AT 1547 EDT ON 6/16/2022 * * *

The following information was received via email: This event is being reported pursuant to 10 CFR 50.72 (b)(3)(xiii) only for a major loss of emergency assessment capability due to the inability to assess primary containment identified and unidentified leakage rates. The containment isolation was not due to actual plant conditions or parameters meeting design criteria for containment isolation. Therefore, this is considered an invalid actuation. Updated ENS Text: During thermography of a reactor protection system (RPS) distribution panel, a circuit breaker (RPS-CB-7B) was inadvertently opened. This resulted in a partial loss of power to RPS Division B, which caused containment isolations to occur in multiple systems (Reactor Water Clean Up, Equipment Drains Radioactive, Floor Drains Radioactive, Reactor Recirculation, and Traversing Incore Probe). Specifically, RWCU-V-1, FDR-V-3, EDR-V-19, RRC-V-19, and TIP-V-15 all closed. All actuations occurred as designed upon the partial loss of RPS power. This is being reported pursuant to 10 CFR 50.72 (b)(3)(xiii) for a major loss of emergency assessment capability due to the inability to assess primary containment identified and unidentified leakage rates. Emergency assessment capability was restored at 1008 PDT upon system restoration. The plant is stable, and all effected systems have been restored. There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified. Notified R4DO (Azua).

• * * UPDATE FROM TRACY HOWARD TO ERNEST WEST AT 1853 EDT ON 8/10/2022 * * *

The following information was received via email: At 0923 (PDT) on June 13, 2022, a partial loss of power to the Reactor Protection System (RPS) 'B' occurred due to the inadvertent opening of circuit breaker RPS-CB-7B during thermography of RPS-PP-C72/P001. The partial loss of RPS 'B' resulted in closure of primary containment isolation valves (PCIVs) in multiple systems. No plant parameters existed which would cause the opening of RPS-CB-7B or actuation of the primary containment isolation; therefore, this is considered to be an invalid actuation of a system listed in 10 CFR 50.73(a)(iv)(B). The closure of PCIVs were expected responses to the partial loss of RPS 'B'. Circuit breaker RPS-CB-7B was closed lo restore energy lo RPS 'B' at 1008 (PDT), containment isolation valves were opened, and the affected systems were returned to normal operating conditions for the current configuration per plant procedures. As indicated in 10 CFR 50.73(a)(1), in the case of an invalid actuation reported under 10 CFR 50.73(a)(2)(iv)(A), the licensee may, at its option, provide a telephonic notification lo the NRC Operations Center within 60 days of discovery of the event instead of submitting a written Licensee Event Report. This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) for invalid actuations reported under 10 CFR 50.73 (a)(2)(iv)(A). This actuation was invalid since it was caused by maintenance activities and not the result of actual plant conditions warranting containment isolation. The following additional information is provided as specified in NUREG-1022: The following inboard containment isolation valves were actuated when personnel inadvertently bumped into RPS-CB-7B during the removal of a panel � RWCU-V-1 Reactor Water Cleanup Suction Inboard Isolation Valve � EDR-V-19 Drywell Equipment Drain Inboard Isolation Valve � FDR-V-3 Drywell Floor Drain Inboard Isolation Valve � RRC-V-19 Reactor Water Sample Inboard Isolation Valve � TIP-V-15 Traversing In-Core Probe Purge Isolation Valve All actuations occurred as designed upon the partial loss of RPS power. There were no actual safety consequences associated with this event since all affected equipment responded as designed. The NRG Residents have been notified. Notified R4DO (O'Keefe).

EN Revision Text: TEMPORARY PROCESS RADIATION MONITORING SAMPLE CART NON FUNCTIONAL At 1245 PST, on November 23, 2019, the Turbine Building Process Radiation Monitoring Sample Rack (TEA-SR-26) was declared non-functional and taken out of service to perform planned preventive maintenance per procedure. The temporary sample cart was placed in service as an alternate method per plant procedures. At 0854 PST, on November 25, 2019, it was discovered that the temporary sample cart had a broken belt. At that time neither the Turbine Building Process Radiation Monitoring Sample Rack nor the temporary sample cart could be returned to service. At 1300 PST, on November 25, 2019, the temporary sample cart was returned to service following repairs. This restored the required compensatory measures for TEA-SR-26. This event is being reported as a major loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

• * * UPDATE ON 11/27/2019 AT 0655 EST FROM SEAN KEEHN TO BRIAN LIN * * *

At 2057 PST, on November 26, 2019, it was discovered that the temporary sample cart had lost power and was not in service. At this time, neither TEA-SR-26 nor the temporary sample cart were in service, this was a subsequent failure of the temporary sample cart, and at the time the station had been unsuccessful at restoring a reliable alternate sampling method following the failure that occurred at 0854 PST, on November 25, 2019. At 2350 PST, on November 26, 2019, the temporary sample cart was returned to service following repairs. This restored the required compensatory measures for TEA-SR-26. The NRC Resident Inspector will be notified. Notified the R4DO (Pick) via email.

On December 18, 2016 at 2320 (PST), a leak was discovered on the High Pressure Core Spray (HPCS) system minimum flow line. The leak is located at a bolted flange downstream of the manual isolation valve HPCS-V-53. The location of the leak is not isolable from the suppression pool. This provides a direct path from inside the Primary Containment to the Reactor Building. High Pressure Core Spray system is a single train Emergency Core Cooling System (ECCS) system, therefore inoperability is reportable per 10 CFR 50.72(b)(3)(v)(D). Based on the location of the leak, Primary Containment integrity is compromised. Primary Containment was declared inoperable and is reportable per 10 CFR 50.72(b)(3)(ii)(A). The cause of the leak is under investigation. Actions are underway to cool down and enter MODE 4. The licensee has notified the NRC Resident Inspector.

• * * RETRACTION FROM MATT HUMMER TO HOWIE CROUCH AT 2245 EDT ON 5/24/17 * * *

Engineering evaluations indicate that there was neither a High Pressure Core Spray (HPCS) system inoperability nor a condition that resulted in a significantly degraded principal safety barrier (Primary Containment). Therefore, this event does not meet the reporting criteria in 10 CFR 50.72(b)(3)(v)(D) and 10 CFR 50.72(b)(3)(ii)(A), and Event Notification# 52443 is being retracted. Bases for the retraction are: (1) Extent or accumulation of water flooding the HPCS room would not have prevented the system from fulfilling any of its designated safety functions, if the system had received a starting signal due to an emergency; and (2) the consequences of the HPCS Minimum Flow Line leak into the Reactor Building were within the dose limits and did not have a significant effect on Primary Containment integrity; therefore, the Primary Containment was degraded but operable. The licensee has notified the NRC Resident Inspector. Notified R4DO (Groom).

Reactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge for approximately 2 minutes during a planned surveillance test due to a subsequent failure of REA-FN-1A (Exhaust Fan) to manually start during restoration from the surveillance test. This event is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. Prior to taking test data the surveillance test directs declaring Secondary Containment inoperable in anticipation of potentially exceeding 0.25 inches vacuum water gauge reactor building pressure during the conduct of the surveillance. Consequently Technical Specification LCO 3.6.4.1.A was entered with a 4 hour completion time to restore Secondary Containment to an operable state. Upon failure of REA-FN-1A to start immediate actions were taken to close reactor building ventilation dampers and secure ROA-FN-1A (Supply Fan). Following closure of ventilation dampers and stopping ROA-FN-1A reactor building pressure was quickly restored to less than 0.25 inches vacuum water gauge with Standby Gas Treatment that was already in operation as part of the surveillance test. There were no radiological releases associated with the event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector. Maximum Secondary Containment pressure noted was 0.1 inches positive water gage.

• * * RETRACTION AT 1351 EDT ON 8/25/2015 FROM MATT HUMMER TO MARK ABRAMOVITZ * * *

Subsequent to the initial report, Columbia has since determined that per NUREG-1022 3.2.7 the event was not reportable as Secondary Containment was 'declared inoperable as a part of a planned evolution ... in accordance with an approved procedure and (Columbia's) TS (Technical Specifications).' No condition has been discovered that would have resulted in the system being declared inoperable prior to the surveillance. Therefore, this event is not considered to be a condition that could have prevented fulfillment of a safety function or a condition prohibited by TS and is not reportable to the NRC as a Licensee Event Report (LEA) per 10 CFR 50.73. The NRC Senior Resident Inspector will be notified. Notified the R4DO (Campbell).